The present invention relates to a method for recovering lithium tritide and subsequently tritium from molten lithium metal. Although the method has application for the recovery of tritium from a blanket of lithium metal exposed to the neutron flux of a thermonuclear reactor, the method can also be used for the recovery of tritium bred within liquid lithium by more conventional means. For example, lithium metal exposed to the moderated neutron flux of a conventional fission reactor employing U-233, U-235 or Pu-239 will be found to include small concentrations of lithium tritide from the tritium bred by neutron reaction with lithium atoms. In either a thermonuclear or a fission reactor a molten lithium blanket could be part of the heat transfer system used in the production of power or merely a blanket for the breeding of tritium. Tritium recovered with the present process can be used as a fuel in a power plant employing the fusion of deuterium and tritium to produce helium and neutrons with a release of energy as in the reaction T(D,n).sup.4 He. Tritium also has application as a radioactive substitute for ordinary hydrogen or protium in biological or other tracer type studies.
Since tritium is unavailable in any significant quantities in nature, a practical and efficient thermonuclear power plant will require the production of tritium for fuel at least at the same rate as it is consumed. A single lithium-6 atom exposed to thermal neutron flux can fission to produce a triton and a helium nucleus. Lithium-7 can be used to offset neutron losses as it will likewise fission on capturing a fast neutron to produce a triton, an .alpha. particle and a neutron. However, lithium-7 has a considerably smaller cross section for low-energy neutron flux than lithium-6.
Another method of compensating for neutron losses is to include beryllium-9 within the lithium blanket or in a separate blanket adjacent to the source of neutrons. Beryllium acts to double the neutron flux in accordance with the reaction .sup.9 Be(n,2n)2.sup.4 He. Where a double blanket system is used, the high-energy neutrons (14 MeV) produced by the deuterium-tritium fusion will first enter a blanket of beryllium where the above reaction will take place, producing about twice the original neutron flux but at a substantially lower energy. A second blanket of lithium will receive the amplified neutron flux werein lithium-6 and lithium-7 isotopes, each capturing a single neutron, produce helium and tritium.
In operating a D-T fueled thermonuclear power plant, it is desirable that only small amounts of tritium be allowed to build up within the lithium blanket before the recovery rate matches the breeding rate. Otherwise, a substantial tritium inventory is required for the startup of the power plant and substantial tritium reserve is contained during operation. The partial pressure of tritium will likewise increase with concentration, thus increasing the rate of tritium permeation through the blanket and heat exchanger structures. Large quantities of tritium in the blanket system will also increase the radioactivity hazards during routine maintenance and emergencies associated with mechanical and structural failures.
It is expected that the lithium blanket of a typical fusion power plant will require a tritium inventory in excess of 10.sup.6 curies or about 2,000 gm of LiT, and it is likely that the permissible level for tritium release to the environment for such plants will be limited to about 1,000 curies/year. This corresponds to a daily loss rate of about 3 ppm or less of the total blanket inventory. Based on anticipated standards, tritium releases in excess of the above figure could begin to constitute an objectionable radiologic hazard. Consistent with these standards, a goal of preferably 10.sup..sup.-10 Torr but no more than 10.sup..sup.-8 Torr partial pressure of tritium over the molten lithium is accepted as a design criteria.
Various processes have been evaluated as methods of recovering tritium from the blankets or coolant systems within fusion reactors. (See Watson, "An Evaluation of Methods for Recovering Tritium from he Blankets or Coolant Systems of Fusion Reactors", U.S. AEC report ORNL-TM-3794, December 1972.) Of these processes, gas sparging, distillation, cold traps and evaporation are thought to be unlikely to reduce the tritium concentration in a lithium blanket to acceptable levels (below about 10 parts per million). Another more promising process involves the use of a permeable window with one surface exposed to the lithium blanket and the opposite surface exposed to an oxygen gas for the formation of T.sub.2 O, but it too includes uncertainties. One possibility for such a permeable window system comprises palladium-coated niobium tubes. However, the ability to make reliable coatings of this nature has not been demonstrated. In addition, it has not been proven how a sufficiently low pressure of tritium can be provided on the metal window surface opposite the lithium blanket to reduce the tritium concentration within the blanket to an acceptable level.
In view of these problems and uncertainties with previously considered processes, the inventor has developed a process for the extraction of lithium tritide into a molten salt. Tritium is thereafter recovered by electrolysis or chemical reaction of the LiT and collection of the evolved tritium gas.